US3146064A

US3146064A - Decontamination of uranium

Info

Publication number: US3146064A
Application number: US307181A
Authority: US
Inventors: Robert L Moore
Original assignee: Individual
Current assignee: Individual
Priority date: 1952-08-29
Filing date: 1952-08-29
Publication date: 1964-08-25
Anticipated expiration: 1981-08-25

Description

3,145,064 DECGNTAMINATION F URANIUM Robert L. Moore, Richland, Wash, assignor to the United States of America as represented by the United States Atomic Energy Commission No Drawing. Filed Aug. 29, 1952., Ser. No. 307,181 Claims. (1. 2314.5)

The present invention is concerned with recovery of uranium and more particularly with the decontamination of uranium values in the course of recovering said values. The present invention is specifically concerned with the decontamination of uranium derived from the recovery processes of neutron-irradiated uranium and with the recovery of uranium from ores.

In the processing of neutron-irradiated slugs aqueous acid solutions are obtained which, in addition to the uranium, contain plutonium and rare earth metal values, the so-called fission products. Separation of the uranium and plutonium is often effected by extraction with an organic water-immiscible solvent whereby the uranium and plutonium are taken up by the solvent while the fission products preferentially remain in the aqueous solution.

Zirconium, one of the fission products, however, has been causing some difficulties, because it is extracted together with the uranium to a substantial degree. Thus the organic extracts as Well as the aqueous rafiinates remaining after extraction usually contain zirconium, and also plutonium and uranium, and it is most desirable to separate and recover these values from the various solutions obtained during the extraction processes.

Likewise, the waste solutions obtained in the extraction processes mostly still contain minor quantities of uranium together with the fission products, and it has also been tried to recover the uranium from these waste solutions.

In order to separate plutonium and uranium from each other by solvent extraction, it has been held necessary heretofore selectively to reduce the plutonium to its trivalent non-extractable state. Ferrous sulfarnate has been used as the reducing agent for this purpose, and the results obtained therewith were highly satisfactory. However, ferrous sulfamate has two distinct drawbacks, name- 1y, it is a rather expensive chemical and it is very unstable.

In the recovery of uranium from ores, solutions are frequently obtained which contain zirconium together with uranium so that the above problem of separating the two co-extractable elements exists also in this instance. Monazite sand, for instance, is one of the ores in which uranium and zirconium coincide.

It is an object of this invention to provide a process for the separation and recovery of uranium by solvent extraction from solutions containing uranium together with zirconium and/or plutonium values.

It is another object of this invention to provide a process for the separation of uranium values from plutonium values by solvent extraction without the necessity of reducing the plutonium to its trivalent state prior to extraction.

It is a further object of this invention to provide a process for reducing the extractability of plutonium and zirconium values by means of organic solvents without thereby adversely aifecting the extraction of uranium values associated with said plutonium and zirconium values.

Other objects and advantages of the present invention will be apparent upon further examination of this specification.

I have discovered that both tetravalent plutonium and zirconium can be converted to a solvent-nonextractable 3.14am Patented Aug. 25, 1964 state by complexing them with a fluoride-containing anion (hereinafter referred to as lino-anion). In other words, uranium contained in mixtures comprising, in addition to uranium, tetravalent plutonium and/or zirconium values can be decontaminated with respect to these values by solvent extraction processes if to an acidic aqueous solution of the mixture a water-soluble fluoanion-containing substance is added prior to the solvent extraction. By contacting the solution then with a substantially water-immiscible organic solvent, the uranium values are taken up by said solvent, while said other metal values remain in aqueous solution.

Fluo-anion-containing compounds found suitable for complexing tetravalent plutonium and zirconium values are the following compounds listed in the descending order of their relative efiiciency for this purpose: sodium fluoride (NaF), ammonium fluosilicate (NH SiF sodium fluosilicate (Na SiF potassium fluosilicate (K SlF or mixtures thereof.

The acidity of the aqueous solution to be treated is advantageously adjusted to between 2.0 and 7.0 M; nitric acid is preferred. Extraction of zirconium by the organic solvent increases with increasing acidity, and a better decontamination of uranium from zirconium is therefore obtained at lower acidities. The preferred acidity range is between 3.0 and 5.0 M.

The fluo-complexing agent is suitably present in concentrations of about 0.01 M for an aqueous solution 3 M in nitric acid; however, the concentration may be as high as 0.1 M. The concentration of the uranium salt may vary widely; however, from 0.1 M to 0.4 M is the preferred range for uranyl nitrate hexahydrate. Zirconium extraction decreases with an increase of the uranium concentration in the solvent.

Normally liquid organic compounds satisfactory for extracting uranium values from the above-described aqueous solutions containing complexing agents for the zirconium values and tetravalent plutonium values pertain to the following classes: ethers, esters, ketones, alcohols, polyethers, alkyl phosphates and alkyl sulfides which are substantially immiscible with water and aqueous solutions. In particular, the following compounds have given satisfactory results in the process of this invention:

Ethyl ether Isopropyl ether Butoxyethoxyethane (ethyl butyl Cellosolve) Diethyl ether of ethylene glycol (diethyl Cellosolve) Dibutyl ether of diethylene glycol (dibutyl Carbitol) Dibutyl ether of tetraethylene glycol Ethyl acetate n-Propyl acetate 7 Butoxyethoxyethyl acetate (butyl Carbitol acetate) Methyl isobutyl ketone (hexone) Acetophenone Mesityl oxide Cyclohexanone Tert-amyl alcohol 2-ethyl-1-hexanol Tributyl phosphate Trioctyl phosphate Dioctyl hydrogen phosphate Octadecyl dihydrogen phosphate n-Propyl sulfide Methyl isobutyl ketone and alkyl phosphates, in particular tributyl phosphate, however, are the preferred solvents.

Extraction is improved by adding a salting-out agent to the aqueous solution to be processed. The preferred salting-out agents for uranium in solvent extraction of uranium with tributyl phosphate from nitric acid solutions are nitrates; the nitric acid itself also acts as a saltingout agent. Generally, the salting-out agent is advantageously a water-soluble salt which has the same anion as the salt to be extracted.

The present invention lends itself also to the separa tion of uranium values from organic solutions containing the uranium together with plutonium and/or zirconium values. In this case the organic solution is contacted with an aqueous solution of fluo-anion-containing substance whereby any plutonium and zirconium are complexed and extracted into the aqueous solution but the uranium is left in the organic solvent.

I have further discovered that in the aqueous solutions the fluo-complexed plutonium and/or zirconium values can be restored to their preferentially organic soluble form by the addition of an aqueous solution of mineral acid or mineral acid salt, advantageously of aluminum nitrate, in a concentration of from 1 to 3 M and preferably of about 1 M, whereby the complex formed of the plutonium and/or zirconium is decomposed. Mixtures of aluminum nitrate and alkali and/or alkaline earth nitrates are also suitable for this purpose. Thereafter the values may be again extracted by an organic solvent and further separation accomplished by repetition of the extraction with a fluo-complexing agent.

The utility of the embodiments outlined hereinbefore and considered either alone or in combination has been confirmed on both a laboratory scale and a plant scale as shown by the accompanying examples.

EXAMPLE I A solution, derived from processing neutron-irradiated uranium, 4.55 M in HNO and containing 93.6 g. per liter of uranium was contacted with a solvent mixture containing 15% by volume of tributyl phosphate and 85% of an inert hydrocarbon having a boiling range near that of kerosene in order to extract and separate the uranium and plutonium from the fission products. The solvent mixture had been previously washed with 1 M sodium hydroxide and water. Nitric acid of a concentration of 6 N was used as scrub solution. The aqueous waste solution, the feed, was introduced into a 14 high and 1" wide extraction column which was packed with A by A Raschig rings. The flow rates of tributyl phosphate mixture and aqueous feed had a ratio of :2. In the solvent phase 1 mg. of uranium had a fi-activity of cts./min. as compared with 15,000 cts./min./mg. of uranium in the aqueous feed solution.

The solvent phase containing the uranium and plutonium was then treated in a second column of similar dimensions for the back-extraction of plutonium according to the process of this invention. The strip solution was an aqueous solution containing 1 g. of ammonium fiuosilicate per liter and nitric acid in a concentration of 2 M. Fresh solvent mixture as used for the extraction in the first column was used as scrub solution. The flow ratio of the aqueous streamzorganic feedzorganic scrub was 1:5: 1. The plutonium content of the uranium in the organic feed solution was reduced by the backextraction from 6.4 10 cts./min./mg. of uranium to 12.5 cts./min./rng. of uranium.

EXAMPLE II Variations in the extraction coefficients (E for plutonium (organic/aqueous) were studied using aqueous tetravalent plutonium solutions 3 M in nitric acid and fluosilicate present as ammonium fiuosilicate in amounts varying between 0.0001 M up to 0.5 M. The foregoing aqueous solutions were contacted at 25 C. with equal volumes of an organic solvent consisting of by volume of vacuum-distilled tributyl phosphate diluted witha hydrocarbon petroleum fraction whose boiling point is in the kerosene range. The results are tabulated below.

Table I Conan. 01(NH4)2 SiFu Ek R" 5 Pu(IV) (g-lh) 15 *R=4.63/Ea.

EXAMPLE III The fiuosilicate does not cause any appreciable complexing of uranium values. Two parallel experiments 20 were carried out, each using identical conditions, and an aqueous feed solution 0.060 M in uranyl nitrate hexahydrate, 0.144 M in H PO 0.121 M in H 80 1.77 M in NaNO and 3.0 M in HNO however, while in one instance the extraction was carried out with this solution as is, in the other instance 1 g. of sodium fluosilicate was added to 1 liter of the solution. After each batch extraction a sample each of aqueous and organic phase was analyzed for uranium. The results are shown in the following table. T able II EFFECT OF FLUOSILICATE ON EXTRACTION OF U (VI) [Equal volume contactings, 25 O.15% TBP-S5% diluent (same as in previous experiment)] 5 Absence of Fluosilicate Ions 1 g./1. Sodium Fluosilicate Batch Extn. N0. Aq. 0 Percent Aq. 0 Percent UNH, En Initial UN H, E Initial g./l. UNH Reg./l. UNH Remaining maining 1 Average.

EXAMPLE IV As shown by the following experimental data, the amount of organic soluble plutonium is dependent on the nitric acid concentration in the feed solution.

Table III VARIATION OF FLUOSILICATE COMPLEXING WITH NITRIC ACID CONCENTRATION [Equal phase eontactings; temperature=25.0 C.Organic: 15% vacuum distilled TBP-85% of the diluent used in Example II] Ea" with 0.01 M M HNO; E,, Amri imi- EL." in absenceof SiFu um uo- '0- I, 60 silicate En 111 presence of 0.01 M SiIa Present The extraction of plutonium into the organic solvent phase appears to be substantially restricted at lower acid 70 concentrations and in the presence of the added fiuosilicate ions.

EXAMPLE V A distribution ratio for tetravalent plutonium of 4.16

was determined after an aqueous Pu (IV)-containing solu tion which was 3 M in nitric acid was contacted with an organic solvent mixture consisting of 15% by volume of tributyl phosphate and 85% of carbon tetrachloride. Thereafter sodium fluoride was added to the aqueous phase in a quantity to obtain a concentration thereof of 0.01 M (the nitric acid concentration had been reduced to 2.94 M by the fluoride addition). The distribution ratio for plutonium was again determined and found to be 0.109. Then sufiicient aluminum nitrate nonahydrate was added to the foregoing sodium fiuoride-complexed aqueous solution to obtain a concentration of 0.092 M of the aluminum nitrate; this reduced the nitric acid concentration to 2.78 M and that of the sodium fluoride to 0.01 M. The distribution ratio for plutonium was again determined and found to be 3.20.

From the foregoing experiments it is apparent that the normally organic soluble tetravalent plutonium is complexed by the fluoride to a preferentially aqueous-extractable form and that this complex can be decomposed and the plutonium thereby restored to the organic solventsoluble form by the addition of aluminum nitrate.

The effect of fluo-anions on the extractability is illustrated in the following example.

EXAMPLE VI An aqueous solution 0.01 M in uranyl nitrate (obtained from neutron-irradiated uranium), and containing 3.4 10 cts./min./ml. of zirconium tracer was contacted for minutes with an equal volume of 0.4 M tributyl phosphate in methylcyclohexane diluent. The two phases obtained were separated and each phase was analyzed for beta-activity and uranium. The results are compiled in Table IV.

From the foregoing data, it is apparent that the uranium ratio is not significantly altered by the complexing agents whereas the zirconium ratio is decreased by a factor of from 12 to 24.

EXAMPLE VII This example shows that, while normally the extraction of zirconium is substantially improved with decreasing uranium content of the organic solution (which again is dependent on that of the aqueous solution), this increase is only nominal when a complexing agent is added according to this invention. Data of comparative experiments are summarized in Table V.

Table V EFFECT OF TBP SOLVENT URANIUDJ CONCENTRATION AND AQUEOUS SOLUBLE COIVIPLEXING AGENTS ON FISSION PRODUCT EXTRACTION Aqueous solution: 3 M nitric acid. Solvent: 001 0.4 M in TBP. Activity: QO-day-cooled slug activity. Equal volumes of aqueous solution and solvent were used for each extraction] Equilibrium Activity Cgmplexing Rance Aqueous U Solvent U gent Comm (M) Comm (M) Aqueous Phase Zr 0. 14.4 0. 162 None .13. 0 0. 0687 0. 132 None 25. O 0. 0175 0. 081 None 67. 0 0. 0065 0. 048 None 103. 0 0. 1424 0. 1624 0. 01 M NazSiF 2. 9(?) 0. 070 0. 1392 0. 01 1VI Na2SiF 1. 6 0. 0173 0. 086 0. 01 M NEmSiF 3. 0 0. 0061 0. 046 0. 01 M NazSiF 4. 2

The change of extractabilitydue to decreasing uranium content-of ruthenium, cerium and uranium was found not to be affected by the presence of a fluo-anion complexing agent.

It will be understood that this invention is not to be limited to the details given herein but that it may be modified within the scope of the appended claims.

What is claimed is:

1. A process for separating uranium values from an acidic aqueous solution containing uranium values and at least one compound selected from a metal value group consisting of zirconium values and plutonium values, which comprises adding a water-soluble salt selected from a salt group consisting of sodium fluoride, ammonium fiuosilicate, sodium fluosilicate, and potassium fluosilicate to said acidic aqueous solution, contacting the resultant solution with a substantially Water-immiscible liquid organic solvent and separating an organic solvent phase containing uranium values from an aqueous rafiinate containing said metal values.

2. A process for recovering zirconium values from an organic solution which comprises contacting said solution with an acidic aqueous solution of a salt selected from the group consisting of sodium fluoride, ammonium fluosilicate, sodium fluosilicate and potassium fluosilicate, and separating an aqueous zirconium-containing phase from an organic ratfinate.

3. A process for separating uranium values from contaminants selected from the group consisting of plutonium values and zirconium values, said values being dissolved together with uranium in an acidic aqueous solution, which comprises adding a water-soluble salt of a group consisting of sodium fluoride, ammonium fluosilicate, sodium fluosilicate, and potassium fiuosilicate to said aqueous solution, contacting the solution with a substantially water-immiscible liquid organic solvent, separating an organic solvent phase containing uranium values from an aqueous rafiinate containing said metal values, adding a water soluble aluminum salt to said aqueous rafiinate, contacting said rafiinate with a substantially water-immiscible liquid organic solvent, separating said rafiinate from an organic metal values-containing solution, and recycling said organic solution for further separation.

4. A process for extracting zirconium values complexed with ions selected from the group consisting of fluoride and fiuosilicate anions from an acid-containing aqueous solution comprising adding a water-soluble aluminum salt to said solution, contacting said solution with a substantially water-immiscible organic solvent, and separating an organic zirconium-containing extract phase from an aqueous raflinate.

5. The process of claim 1 wherein the concentration of the water soluble salt in the aqueous phase is from 0.01 M to 0.1 M.

6. The process of claim 1 wherein the organic solvent is an alkyl phosphate.

7. The process of claim 1 wherein the organic solvent is" tributyl phosphate.

8. The process of claim 2 wherein the organic solution contains tributyl phosphate in a concentration of about 0.4 M.

9. The process of claim 2 wherein the aqueous phase is 3 M in nitric acid and 0.01 M in other salt.

10. The process of claim 3 wherein said aluminum salt is an aluminum nitrate and said acid is nitric acid.

References Cited in the file of this patent UNITED STATES PATENTS 2,227,833 Hixson et a1. Jan. 7, 1941

Claims

1. A PROCESS FOR SEPARATING URANIUM VALUES FROM AN ACIDIC AQUEOUS SOLUTION CONTAINING URANIUM VALUES AND AT LEAST ONE COMPOUND SELECTED FROM THE A METAL GROUP CONSISTING OF ZIRCONIUM VALUES AND PLUTONIUM VALUES, WHICH COMPRISES ADDING A WATER-SOLUBLE SALT SELECTED FROM A SALT GROUP CONSISTING OF SODIUM FLUORIDE, AMMONIUM FLUOSILICATE, SODIUM FLUOSILICATE, AND POTASSIUM FLUOSILICATE TO SAID ACIDIC AQUEOUS SOLUTION, CONTACTING THE RESULTANT SOLUTION WITH A SUBSTANTIALLY WATER-IMMISCIBLE FLUID ORGANIC SOLVENT AND SEPARATING AN ORGANIC SOLVENT PHASE CONTAINING URANIUM VALUES FROM AN AQUEOUS RAFFINATE CONTAINING SAID METAL VALUES.